Text

Amend 15 to License NPF-86,changing Plant TS to Allow Relaxation in Psv & MSSV Setpoint Tolerances to +3%
	ML20114D965
Person / Time
Site:	Seabrook
Issue date:	09/03/1992
From:	Butler W; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20114D966	List: ML20114D965; ML20114D967;
References
	NUDOCS 9209100228
	Download: ML20114D965 (11)
	v • d • e

.

/"

UNITED STAT [s

/*

"?,

NUCLE AR REGULATORY COMMISSION 3.,.

[

j WASHINGTON. D C 20tn

's, 1

/

NORTH ATLANTIC ENERQY SERVICE CORPORATION. ET AL.*

DOCKET NO. 50-443 SEABR00KJ Tall 0N. UNIT NQm_1 M[fiDMENT TO FACILITY OPERAllNG LICENSE Amendment No15 License No. NPF-86 1.

The Nuclear Regulatory Comis'lon (the Comission or the NRC) has found that:

A.

The application for amendment filed by the North Atlantic Energy Service Corporation (NAESCO) (the licensee), acting P-itself and as agent and representative of the 11 other utilities sisted below and hereaf ter referred to as licensees, dated May 5,1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will opeiate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance:

(1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and; E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

' North Atlantic Energy Servir.e Corporation is authorized to act as agent for the North Atlantic Energy Corporation, the Canal Electric Company, The Connecticut Light and Power Company, EUA Power Corporation, Hudson Light &

Power Department, Massachusetts Municipal Wholesale Electric Company, Montaup Electric Company, New England Po m Company, New Hampshire Electric Cooperative, Inc., Taunton Municipal Light Plant, The United 111uminating Company, anj Vermont Electric Generation and Transmission Cooperative, Inc.,

and has exclusive responsibility and control over the physical construction, operation and maintenance of the facility.

$$h0$!!

f P

2.

Accordingly, the license is arended by changes to the Technical Specifications as indicated 1 the attatriment to this license amendment, and paragraph 2.C.(2) of fact, y Operating License No. NPF-86 is hereby amended to read as follows:

(2)

Technical Soecificationi The Technical Specifics; ions contained in Appendix A, as revised through Amendment No.15, and the Environmental Protection Plan contained in Appendix B are incorporated into facility License No.

NPF-86. NAESCO shall operate the facility in accordance with the Technical Spet.ifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days of receipt of this letter.

FOR THE NVCLEAR REGULATORY COMMISSION

/acA? A Walter Butler Director Project Directorate 1-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: September 3, 1992

- - - - - - - - - - - - - - - - - - - - - - - ~ -

i ATTACHMENT TO LICENSL AMENDMENT NO.15

[3fillTY OPERATING LICENSE NO. NPF-86 DOCKET NO. 50-443 Replace the following pages of the Appendix A Technical Specifications with the attached pages.

The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

0verlap pages have been provided.

fletgyt iniert 3/4 4-7*

3/4 4-7*

3/4 4-8 3/4 4-8 3/4 4-9 3/4 4-9 3/4 4-10*

3/4 4-10*

3/4 7-l*

C 3/4 7-l*

3/4 7-2 3/4 7-2 B 3/4 7-1 B 3/4 7-1 B 3/4 7-2*

B 3/4 7-2*

RfAUl0LC00LANT$4 STEM REACTOR COOLANT LOOPS AND COOLANT CIRCULATION COLD SNUTDOWN - LOOPS NOT FILLED LIMITING CONDITION FOR OPERATION 3.4.1.4.2 Two residual heat removal (RHR) loops shall be OPERABLE

and at least one RHR loop shall be in operation.**

APPLICABILITY:

MODE 5 with reactor coolant loops not filled.

ACTION:

With less than the above required RHR loops OPERABLE immediately a.

initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible, b.

With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Rer.c+.or Coolant System and immediately initiate corrective action to tri. urn the required RHR loop to operation.

SURVEILLANCE REQUIREMENTS

~

4.4.1.4.2 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

I l

0ne RHR loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other RHR loop is OPERABLE and in operation.

l-

- The RHR pump may be deenergized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided:

(1) no opera-tions are permitted that would cause dilution of the Reactor Coolant System boron concentration and (2) core outlet temperature is maintained at least j'

10*F below saturation temperature.

SEA 8R00K - UNIT 1 3/4 4-7 I

i-d l

e-w-

-r, r

-, +,

e----

-. - - -.., ~. - -

,,,. - ~ -,

r,--

-.m.

REAC70R COOLANT SYSTEM 3/4.4.2 SAFETY VALVES SHUT 00WN LIMITING CONDITION FOR OPERATION 3.4.2.1 A minimum of one pressurizer Code safety valve shall be OPERABLE with a lift setting

of 2485 psig i 3%.**

APPLICABILITY:

H0 DES 4 and 5.

ACTION:

With no pretsurizer Code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode.

i SURVEILLANCE REQUIREMENTS 4.4.2.1 No additional requirements other than those required by Specifica-tion 4.0.5.

The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

- Within i 1% following pressurizer Code safety valve testing

.SEABROOK - UNIT 1 3/4 4-8 Amendment No.15

REACTOR C0OLANT SYSTEM SAFETY VALVES

[

. OPERATING r

LIMITING CONDITION FOR OPERATION 3.4.2.2 All pressurizer Code safety valves shall be OPERABLE with a lift setting

of 2485 psig

3%.**

APPLICABILITY:

MODES 1, 2, and 3.

ACTION:

With one pressurizer PMe safety valve inoperable, either restore the inoper-able valve to OPERABLt statue within 15 minutes or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.2.2 No additional requirements other than those required by Specifica-tion 4.0.5.

The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

- Within

1% following pressurizer Code safety valve testing

ntry into this M00E'is permitted for up to U hours'to perform post-E modification or post-saintenance testing to verf ?y OPF.RABILITY of components.

ACTION requirements shall not apply untti UPER4BILIT1 bas tren verified.

SEABROOK - UNIT 1 3/4 4-9 Ameaent No.15

... 2....

REACTOR COOLANT SYSTEN 3/4.4.3 PRESSURI2ER LiHITING CONDITION FOR OPERATION 3.4.3 The pressurizer shall be OPERABLE with a water volume of less than or equal to 92% of pressurizer level (1656 cubic feet), and at least two groups of pressurizer heaters each having a capacity of at least.50 kW.

APPLICABILITY:

MODES 1, 2, and 3.

ACTION:

With only one group of pressurizer heaters OPERABLE, restore at least a.

two groups to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With the pressurizer otherwise inoperable, be in at least HOT'5TANOBY with the Reactor Trip System breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.3.1 The pressurizer water volume shall be determined to be within its limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.3.2 The capacity of each of the above requirev groups of pressurizer heaters shall be verified by energizing the heaters from the emergency power supply and measuring circuit current at least once per 92 days, i

SEABROOK - UNIT 1 3/4 4-10

-...-. ~~

3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION

3. 7.1.1 All main steam line Code safety valves associated with each steam gsnerator shall be OPERABLE with lif t settings as specified in Table 3.7-2.

i APPL!t.'6ILITY:

MODES 1, 2, and 3.

ACTION:

With four reacior coolant loops and associated steam generators in operation and with one or mere main steam line Code safety valves inoperable, operation in H0 DES 1, 2, and 3 may proceed. provided that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either the in-operable valve is restered to OPERABLE status or the Power Range Neutron Flux

~,

High Trip Setpoint is reduced per Table 3.7-1; otherwise, be in at-least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional requirements other than those required by Specifica-cion 4.0.5.

f i

I

ntry;into this MODE is permitted for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to perform post-i E

. modification or post-maintenance testing to verify OPERABILITY of components.

ACTION-requirements shall not apply until OPERABILITY has hein verified.

SEABROOK - UNIT 1 3/4 7-1

.w..-.-,

w,

,w--.

.c m

w e rmr-

1 TABLE 3.7-1 e

KAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH IkOPERABLE $ TEAM LINE SAFEf7' VALVES OUAING FOUR-LOOP OPERATION MAXIMUM NUMBER OF INOPERABLE MAXIMMM ALLOWABLE POWER RANGE SAFETY VALVES ON ANY NEUTRON FLUX HIGH SETPOINT OPERATING STEAM GENERATOR,

(PERCENT OF RATED THERRAL POWER) 1 87 2

65 3

43 TABLE 3.7-2 STEAM LINE SAFETY VALVES PER LOOP VALVE NUMBER Loop 1 Loop 2 Loop 3 Loop 4 LIFT SETTING * (* 3%)**

ORIFICE SIZE V6 V22 V36 V50 1185 psig 16.0 sq. in.

V7 V23 V37 V51 1203 psig 16.0 sq. in.

V8 V24 V38 V52 1220 psig 16.0 sq. in.

V9 V25 V39 V53 1238 psig 16.0 sq. in.

V10 V26 V40 V54 1255 psig 16.0 sq. in.

The lift setting pressure shall correspond to ambient conditions of the valve

.at nominal operating temperature and pressure.

- Within i 1% following r, sin steam line Code safety valve testing s

SEABROOK - UNIT 1 3/4 7-2 knendment No.15 1

3/4.7 PLANT SYSTEMS BASES 3/4.7.,

TURBINE CYCLE 3/4.7.1,1 SAFETY VALVES The OPERABILITY of the main steam line Code safety valves ensurer. that the Secondary System pressure will be limited to within 110% (1320 psia) of its design pressure of 1200 psia during the most Severe anticipated system

,perations1 transient.

The maximum relieving cia.7 city is associated with a furbink trip from 100% RATED THERHAL POW ~R coinciJent with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).

The specified valve lift settings and relieving capacities are in accor-dance with the requirements of Section III of the ASME Boiler and Pressure Code, (1974 Luition, including the Summer 1975 Addenda).

The total relieving capacity for all valves on all of the steam lines is 1.839 x 107 lbs/hr which is 121% of the total secondary steam flow of 1.514 x 107 lbs/hr at 100% RATED THERMAL POWEtt.

A minimum of two OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-1.

STAL 'UP and/or POWER OPERATION is allowable with t:'-b valves inoperable within the limitations of the ACTION requirements on the u :, of the reduction in Secondary Coolant System steam flow and THERMAL POWE.9 required b;r the reduced Reactor trip settings of the Power Range Neutron Flux channels.

The Reactor Trip Setpoint reductions are derived on the following bases:

For four loop operation:

3p, (X) - (Y)(V) x 109 X

Where:

SP = Reduced Reactor Trip Setpoint in percent of RATED THERMAL Pot lER, V = Maximum number of inoperable safety valves per steam line, 109 = Power Range Neutron Flux-High Trip Satpoint for four loop operation, X = Total relieving capacity of all safety valves per steam line in 1bs/hr, and Y = Maximum relieving capacity of any one safety valve in i

lbs/hr l

SEABROOK - UNIT 1 B 3/4 7-1 Amendment No.15 l

l

PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE (Continued) 3/4.7.1.2 AUXILIARY FEE 0 WATER SYSTEM The OPERABILITY of the Auxiliary Feedwater System ensures that th( so

,or Coolant System can be cooled down to less than 350*F from normal operats w; conditions in the event of a total loss-of-offsite power.

~

The electric motor-driven emergency feedwater pump is capable of deliver-ing a total feedwater flow of 650 gpm at a pressure of 1221 psig to the en-trance of the steam generators.

The steam-driven emergency feedwater pump is capable of delivering c total feedwater flow of 650 gpm at a pressure of 1221 psig to the entrance of the steam generators.

The startup feedwater pump serves as the third auxiliary feedwater pump and can be manually aligned to be powered from an emergency bus (Bus 5).

The startup feedwater pump is capable of taking suction on the dedicated amergency feedwater volume of water in the condensatt storage tank and delivering a totol feedwater flow of in excess of 650 gpm at a pressure of 1221 psig to the entrance of the steam generator via either che main feedwater header or with manual alignment to the emergency feedwater flow path.

This capacity is sufficient to ensure that adequate feed-water flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350*F when the Residual Heat Removal System may bt placed into operation.

3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water vol-use ensures that sufficient water is available to cool the RCS to a temperature of 350*F.

The OPERABILITY of the concrete enclosure ensures this availability of water following rupture of the condensate storage tank by a tornado generated missile. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

3/4.7.1.4 SPECIFIC ACTIVITY The limitations on Secondary Coolant System specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of a stema line rupture.

This dose also includes the effects of a coincident 1 gpa reactor-to-secondary tube leak in the steam generator of the affected stehm line. These values are consistent with the assumptions used in the safety analyses.

3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blow down in the event of a steam line rupture.

This restriction is required to:

(1) minimize the positive reac-tivity effects of the Reactor Coolant System cooldown associated with the blowdown, and (2) limit the pressure rise within containment in the event the steam line rupture occurs within containmont.

The OPERABILITY of the main steam isolation valves within the closure times of the Surveillance Require-ments are consistent with the assumptions used in the safety analyses.

SEABROOK - UNIT 1 B 3/4 7-2

_

ML20114D965

Text

Attachment:

Navigation menu

Search